The present invention relates to nuclear reactors and, more specifically, to techniques for cooling nuclear reactors, and their containments, in the event of one or more types of malfunctions.
During normal operation of a nuclear reactor, the nuclear fuel in a reactor vessel remains covered with water to generate steam. The nuclear fuel consists of fuel rods which develop substantial internal heat. After shutdown of the reactor, decay reactions continue to generate heat in the fuel rods for a lengthy period.
It is important to environmental safety to ensure that radioactive materials are not released during normal or abnormal operation. Such radioactive materials include, for example, steam generated in the reactor vessel and water that is condensed from such steam.
Abnormal operations include, for example, a loss-of-coolant accident which may occur due to a break in a component or piping such as, for example, a steam pipe, within the containment building. The three requirements in such a situation are (1) to refill the reactor pressure vessel with water to cover the fuel rods, (2) to dissipate the heat existing immediately following the break, and (3) to remove the decay heat over an extended period (days or weeks) following the break, such that structural integrity of the containment vessel is maintained.
In the prior art, the movement of cooling water to satisfy the three foregoing requirements is provided by high-pressure water pumps driven by electricity or other external power source. In the event of failure of the normal electrical grid supplying electric power to the plant, diesel generators are provided to take over the task of supplying power for driving the pumps. It is a fact, however, that there is a small but finite probability that diesel generators can fail to function at a critical time, or that human errors can incapacitate systems. Such failure following a serious loss of coolant accident such as, for example, a break in a steam pipe, can be considered a worst-case scenario.
In the prior art, a suppression pool is disposed in a wetwell within the containment building for a nuclear reactor. The wetwell pool includes a substantial volume of water in the suppression pool as well as a substantial volume of gas. During an accident, steam enters the suppression pool through vents from the drywell wherein the supply of water, initially at a temperature of 100 degrees F., or below, cools and condenses the steam. The design values of the containment permit an internal pressure of up to about 55 psig. This implies a maximum steam saturation temperature of about 248 degrees F. A suppression pool of a practical size is incapable of absorbing all of the heat initially contained in the steam vented to the drywell as well as the additional decay heat that must be dissipated over the ensuing several days, without exceeding the maximum temperature and pressure.